Plan B

“Plan-B”An Alternative Liquidation* Strategy

of Fukushima Daiichi NPP

May 21, 2011Satoshi Sato

[email protected]

International Access Corporation

*: A term “liquidation” is used in this document to generally mean various activities directly and indirectly associated with restoration of safe state of each affected reactor in Fukushima Daiichi NPP. This follows a precedent in which workers involved in the emergency actions on the Chernobyl site during the accident and the subsequent clean-up operations were called “Liquidators”.

2

BWROG BWR Owners Group

CCI Core-Concrete Interaction

EPG Emergency Procedure Guidelines

FHM Fuel Handling Machine

FP Fission Product

GTCC Greater Than Class C (Cask for High Level Radiation Waste)

ISFSI Independent Spent Fuel Storage Installation

NPP Nuclear Power Plant

OHC Overhead Crane

RPV Reactor Pressure Vessel

SFP Spent Fuel Pool

SGTS Standby Gas Treatment System

SNF Spent Nuclear Fuel

Abbreviations

3

4

5

Contents

• Present Status• “Plan-A” Ever discussed?

– Difficulties to dismantle BWR reactor with damaged core

– Difficulties to retrieve spent fuel from degraded Reactor Bldg.

– Conventional Decommissioning Processes– Definition of “Plan-A”

• Developing “Plan-B”– Worst Case Scenario

6

Contents (cont’d)

• “Plan-B”– Liquidation Strategies for Fukushima NPP Reacto

rs and SFPs– Water Treatment and Entombment– On-Site Above-Ground Repository– Design Beyond Millennium– Beyond “Liquidation

• Next Step

7

Present Status

• Reactor Status– Reactor Core

– Reactor Pressure Vessel

– Primary Containment

– Reactor Building (Secondary Containment)

– Residual Heat Generation

• Spent Fuel Pool Status– Fuel Integrity

– Pool

– Residual Heat Generation

8

Summary - Reactor

UnitReactor

CoreReactor

Pressure VesselPrimary

ContainmentReactor Building

1 Totally destroyed

Barrier integrity no longer maintained. Bottom Head Penetrations severely damaged.

Barrier integrity no longer maintained. Details not confirmed.

Original Function as Secondary Containment totally lost due to H2 explosion on Refueling Floor. Remaining part of building still reasonably good.

Overhead Crane (OHC) and Fuel Handling Machine (FHM) not available.

2 Ditto Ditto

Barrier integrity severely degraded due to H2 explosion inside or outside Torus.

Function as Secondary Containment still reasonably maintained even after H2 explosion. OHC and FHM still fully functioning.

3 Ditto Ditto Same as Unit 1Same as Unit 1, except that some portions lower than Refueling Floor also degraded due to H2 explosion

4 Empty Not affected Not affected Ditto

Function Barrier Integrity

Severely damaged Severely damaged

Severely damaged Possibly still partly maintained but not confirmed

9

Unit 4 Unit 3 Unit 2Unit 1

State of Reactor Building, Unit 1 to 4 looking from east as of March 20

10

Unit 3 Unit 4

State of Reactor Building, Unit 3 and 4 looking from west as of March 20

11

Reactor Core, Pressure Vessel, and Primary Containment

• Current Degree of Degradation of each FP Barrier– Melt-Down through Core Plate: No Doubt

– Leakage of RPV Bottom Head: No Doubt

– Gross Failure of RPV Bottom Head: Not Very Likely

– Leakage of Primary Containment: Highly Likely

– Gross Failure of Primary Containment:

• Unit 1 and Unit 3: Not Very Likely

• Unit 2: Highly Likely (Suppression Chamber)

– Major Core-Concrete Interaction （ CCI ） : Not Very Likely

– Melt-Down through Man-made Rock (Basemat): Not Likely

12

12~

13m

~33.5m

ID 8.9m

~46m

~23.5m

~15m

~11m

~40m

16~

17m

Refueling Floor

Rx. Bldg. (Secondary Containment)

Primary Containment

Reactor Pressure Vessel

Suppression Chamber (part of Primary Containment)

Drywell

Pedestal

Typical Configuration　（ Unit 3, 4）

13

Melt-Down through Core Plate

Predicted to occur 2 hours following complete loss of cooling capability.

Several previous experiments suggested steam explosion not likely.

Core Shroud

Core Plate

Reactor Core

Molten Core

Water

No Doubt

14

Further Melt-Down through Core Plate

Unit Duration

1 14h09m

2 06h29m

3 06h43m

Actual Complete Loss of Cooling Capability

(Official Announcement by Government on May 16, 2011)

No Doubt

15

Degradation of Reactor Pressure Vessel Bottom Head

Creep rupture begins to occur at ~240-deg C below melting point (1500-deg C) of vessel material (low alloy steel), allowing some leakage of highly contaminated water containing fractured pieces of fuel pellets.

Highly Likely

16

Locations of Potential Leakage　 (Typ.)

Vulnerability of Bottom Head Leakage

17

Further Degradation of Reactor Pressure Vessel Bottom Head

Drywell Sump Pit

Pedestal

Pedestal DoorwayPossible

18

Major Degradation of Reactor Pressure Vessel Bottom Head and Core-Concrete Interaction (CCI), Resulting in Significant Amount of Release of Radioactive Aerosol

Pedestal Doorway

Pedestal

H2O,

CO2

H2O,

CO2

H2, CO

Aerosol

AerosolAerosol

Aerosol

Not very likely, but could have happened depending on cooling evolution during early stage.

19

Beginning of Primary Containment Failure

Pedestal DoorwayPedestal

Aerosol

Aerosol

Aerosol

Aerosol

H2, CO

H2O,

CO2

H2O,

CO2


20

Pedestal Doorway

Pedestal Wall

Source: NUREG/CR-6042 Rev.2

21

Beginning of Primary Containment Failure

Aeros

ol


22

Not likely

Failure due to Creep Rupture

Gross Failure of Primary Containment due to Steam Explosion

23

Gross Failure of Primary Containment due to Melt-Down

Aeros

ol

Aeros

ol

Aeros

ol

Aeros

ol

Not likely

24

Complete Melt-Down through Man-made Rock (Basemat)

Not likely

Man-made Rock

25

Unit 1 2 3 4

# of Fuel Assembly in Rx. 400 548 548 0

Electrical Output (MWe) 460 784 784 0

Thermal Output (MWt) 1,380 2,381 2,381 0

Estimated Residual Heat (MWt)0.1% of rated Thermal Output

1.4 2.4 2.4 0

Residual Heat Generation

2 months after

shutdown

26

Summary - SFP

• Fuel Integrity– No conclusive information so far.– Potential thermal damage. (Units 3 and 4)– Potential mechanical damage. (Units 1 to 4)

• Pool– Details unknown, but apparently no major damage. – Potential thermal damage due to overheating.

(Units 3 and 4)– Potential mechanical damage due to earthquake and/or H2

explosion.

(Units 1 to 4)

27

Unit 4 SFP

Top view of fuel rack by remote underwater TV camera. Difficult to draw any conclusion about fuel integrity only based on this information. Fuel inspection by “sipping” is warranted.

28

Unit 1 2 3 4

Number of Fuel Assembly

New Fuel Storage Vault 100 28 52 204

Spent Fuel Pool 292 587 514 1331

Hottest Spent Fuel Discharged

(Date of beginning of last refueling outage)3/25/’10 9/16/’10 6/19/’10 11/30/’10

Estimated Residual Heat Generation Rate (MWt) 0.07 0.46 0.23 1.8


29

Intentionally left blank

30

“Plan-A”Has it ever been discussed?

• Efforts to achieve so-called “Cold Shutdown” have been being made as the single topmost priority.– Originally, this term appears to specifically mean a state par

tially submerging the reactor core from both inside and outside Reactor Pressure Vessel (RPV) to keep the entire metal surface temperature below 200-deg F as defined by Tech Spec.

– However, after realizing that the barrier integrity of the RPV has been excessively challenged, this term now means a submersion only from outside RPV.

– In more common wording, “Drywell Flooding”. – Consistent with the intent per BWROG’s EPG.

31

Drywell Flooding

32

Issues – Public Perspectives

• “Cold Shutdown” is apparently considered to be only the first milestone of the entire “Liquidation” program . It is not a solution or a goal. Nothing has been told to the public beyond that point. Both tax-payers and rate-payers are concerned about the cost and schedule of the entire program, and above all, the capability to ultimately get it done.

• “Cold Shutdown” is used as a magic word. Public expects this would drastically improve the radiological environment so that refugees may be able to go home soon after this event. But this will never become true because contamination already there will continue to stay there regardless of the state of reactors.

33

Issue – Technical Perspectives• Applying BWROG’s EPG after having significantly deviated fro

m it currently creates a major “side effect”, that is, generating large volume of highly contaminated water. The longer they operate “Cold Shutdown”, the more contaminated water they generate and partly leak to the external environment (groundwater and seawater).

• Cannot apply it for Unit 2 because of major defect caused on the pressure boundary of its Suppression Chamber.

• Cannot cool the vicinity of “gas pocket” at the top of Skirt. The NRC report (NUREG/CR-6402 Rev.2) had pointed out a possibility of delayed creep rupture.

34

Delayed wall creep rupture would eventually occur in the vicinity of gas pocket.

35

Questions

• Is “Cold Shutdown” a mandatory milestone even 2 months after cease of chain reaction?

• Is there any better approach to manage the residual heat which is now only ~0.1% of the rated thermal power?

• How worse could it be at all if “Cold Shutdown” is abandoned? It has never been achieved to date for a long time anyway. What is the reason to have to stick to it?

• Achieving “Cold Shutdown” is OK. But, then what? – How long to keep it?

– Dismantling reactors, next?

– Back to “Green Field” eventually?

36

Difficulties to Dismantle BWR Reactors with Damaged Core

Reasons why so difficult• Requires full restoration of Refuel Floor, a part of Secondary C

ontainment and Reactor Building itself along with its ventilation system, as well as Overhead Crane, Fuel Handling Machine and all other special service tools prior to disassembly of RPV Head.

• Flooding RPV would result in leakage from the Bottom Head.• Removal of RPV Head is a challenging task.

– Studs/Nuts potentially severely galled.

– Radiation level too high to install/operate Stud Tensioner.

• Removal of Steam Dryer is more challenging due to high radiation level.

37

Reasons why so difficult (cont’d)• Removal of Moisture Separator is even more challenging.

– Shroud Head Bolt possibly galled due to exposure to elevated temperature and cannot be unlatched by following conventional procedure. Possibly requires to sever by remote EDM.

• Removal of Fuel Assembly is yet more challenging if not impossible. Most, if not all, FAs could have been fully destroyed, deformed and fused each other. Most part of core now possibly locates below Core Plate as “core debris”.

• Complete retrieval of once-molten core debris below Core Plate requires an exhaustive effort.

• Once core debris exits RPV and flows into the pedestal region, further efforts are not even theoretically possible unless Primary Containment Vessel is totally flooded.

38

Steam Dryer

39

Moisture Separator

40

Fuel Assembly

41

Lower Plenum (Region below Core Plate)

42

Justification to pursue• Efforts to place nuclear material under better inventory control

are in line with IAEA requirements for security reason.• Efforts to contain nuclear material within certified containers as

much as possible are considered more ethical practice.• Possible contribution to reduce long term risks associated with

decay heat and radio-toxicity. • Opportunity to gain detail technical data to be shared with

international community to improve management of severe accident. (e.g. Improving accuracy of analytical codes.)

Justification not to pursue• Too much technical and financial uncertainties to pursue.• Possibility to reduce short term risks associated with safety

and security. (e.g. Exposing damaged reactors to unsecured condition for extended period not preferable.)

43

Conclusions• Dismantling BWR Reactors with damaged core is technically

extremely challenging due to harsh radiological environment and not fully achievable any way.

• “Liquidation” strategy with damaged reactor left as is should be considered as one of the practical options.

• However, such a shortcut option requires public support domestically (including local communities) and internationally. Concurrence from IAEA may be necessary to comply with requirements for an NPT member country like Japan.

44


45

Difficulties to retrieve spent fuel from degraded Reactor Bldg.

Why difficult?• Currently no place to go and no certified container to load for t

he damaged fuel assemblies. Therefore, fuel inspection (sipping) or other technically acceptable method to separate damaged fuel from intact fuel is necessary step to proceed.

• Fuel Handling Machine (FHM) is needed for fuel inspection.• Overhead Crane (OHC) with a loading capacity greater than 100

MT is needed for lifting spent fuel casks.• Both FHM and OHC are not currently functionable. Restoration

of OHC requires a major work to repair degraded Reactor Bldg.

46

Why difficult? (cont’d)• Using other type of crane or lifting machine is an optional

choice. However, it needs to be certified for safety application when handling spent fuel casks.

• New analyses and/or experiments may be required for the existing cask designs to be used.– All existing cask designs have been certified to meet requirements

to survive a set of design basis accidents including 9-meter free drop. However, using other crane may allow exceeding this condition.

– “Fuel-Handling Accident” (accidentally dropping one fuel assembly over the core) has been assumed as one of the design basis accidents for safety analysis and licensing both in Japan and the US. Standby Gas Treatment System has been assumed available under this postulated accident.

– “Spent Fuel Cask Drop Accident” is another design basis accident considered in the US for licensing, but not in Japan.

47

48

Justification to pursue• Efforts to place nuclear material under better inventory control

are in line with IAEA requirements for security reason.• Efforts to contain nuclear material within certified casks as

much as possible are considered more ethical practice and to be exercised wherever reasonably possible.

• Possible contribution to reduce long term risks associated with decay heat and radio-toxicity.

Justification not to pursue• Too much financial hardship to pursue.• Possibility to reduce short term risks associated with safety

and security. (e.g. Exposing physically unprotected SFP to unsecured condition for extended period not preferable.)

49

Conclusions• Retrieving SNF from degraded Reactor Bldg. is technically

challenging under existing licensing scheme especially when FHM and OHC are not available, and without intact Secondary Containment as well as emergency ventilation system (SGTS).

• “Liquidation” strategy with all SNF left in the existing SFP should be considered as one of the practical options for the units where FHM, OHC, Secondary Containment, and SGTS are lost.

• However, such a shortcut option requires public support domestically (including local communities) and internationally. Concurrence from IAEA may be necessary to comply with requirements for an NPT member country like Japan.

50


51

Conventional Decommissioning Processes

Typical Decommissioning Processes in the US• Experiences:

– PWR: Yankee Rowe, Haddam Neck, Maine Yankee, Trojan

– BWR: Big Rock Point

• Goal: back to “Green Field”• Licensing Process:

– PSDAR (Post Shutdown Decommissioning Activity Report)

– LTP (License Termination Plan)

• Spent Fuel: Loaded in Dry-Cask and stored at site (ISFSI) until federal government determines what to do.

• Major Impacts Previously Experienced:– Inflation of Disposal Cost

– Significant Soil Contamination

52

Maine Yankee Experience• Reactor Type/Size: Three-Loop PWR, 2,700MWt/860MWe• Operational History: 12/28/1972 – 12/06/1996 (24 years)• Decommissioning Activities:

– Facility Demolition: 1997 – 2005

• Original Cost Estimate (as of 1997): $380M

• Overrun Cost: $26.8M

– Removal of Spent Fuel: Aug. 2002 – May 2004

• Original Cost Estimate (as of 1997): $128M

• Overrun Cost: $6.8M

– Personnel Exposure:

• Original NRC estimate: 11,150men-mSv

• Actual Result: Approx. half of estimate

– Techniques Applied:

• Chemical Decontamination

• Implosion

• Underwater High Pressure Abrasive Water Jet Cutting

53

Maine Yankee Experience (cont’d)• Decommissioning Activities (cont’d):

– Radioactive Waste:• Total Amount: ~140,000ton

• Amount transported to storage site: 88,450ton (63% of total) mostly by train

– Radioactive Waste (Reactor Internal Components)

Weight Activity

Ton % Bq %

Loaded and shipped in RPV 116 70 0.15E16 2

Loaded and shipped in casks 33 20 1.09E16 15

Loaded in GTCC and stored at site (4 GTCC casks)

16 10 6.03E16 83

Total 165 100 7.26E16 100

54

86% completion as of April 14, 2004

55

Implosion on Containment Bldg., September 17, 2004

97% completion as of January 19, 2005

56

Essentially 100% completed, as of May 5, 2005

“Green Field” achieved on July 25, 2005

57

Implosion Technique

Applied for Turbine Bldg.

58

ISFSI Pad and Spent Fuel Storage Casks

Vertical

Horizontal

59

Yankee Rowe Experience

ISFSI for Storage of 16 dry casks containing 533 spent fuel assemblies

Prior to Decommissioning Activities (1993)

Most Decommissioning Activities done (12/12/2006)

Decommissioning cost : $608M

600MWt PWR (1963 – 1991)

60

Back to “Green Field” as of 9/5/2007

Actual and Future Yankee Rowe Decommissioning Schedule

61

Cased in container on 11/20/1996

Departed from site on

4/27/1997

Loaded on to railcar for 1800km

transportation

Arrival at Barnwell Site

for subsurface

repository on 5/7/1997

Reactor Vessel Disposal

3.6m-dia. x 8.1m-tall, weighing 165tons

80 tons of concrete poured

inside and outside vessel

62

Large volume of subsurface soil found contaminated with tritium (H-3).

Numbers indicate H-3 concentration in groundwater in pCi/L.

EPA drinkable level is 20,000pCi/L.

63

Zion Project

3250 MWt PWR

Operational History

Unit 1 thru 1996 Unit 2 thru 1997

64

Basically just cooling-down

Dismantling Activities

65

30-year long project!

Finally Back to Green Field in 2028

SNF Disposition Campaign

66Source: NUREG-1350 Vol.21

Ultimate Solution for SNF if not Recycled

Yucca Mountain Project (abandoned) 500 to 600m deep geological repository

6767

Swedish Plan (active)

68


69

Definition of “Plan-A”

• It has never been formally defined yet.

• So let’s assume that it is a recovery process, or decommissioning process to achieve “Green Field” as traditionally attempted in Japan (e.g. JPDR) and for several NPPs in the US.

• It has never been formally defined yet.

• So let’s assume that it is a recovery process, or decommissioning process to achieve “Green Field” as traditionally attempted in Japan (e.g. JPDR) and for several NPPs in the US.

70

Feasibility to apply US Decommissioning Experiences and Lessons-Learned for Fukushima NPP

• Demolition Techniques:– Some applications possible but scope limited.

– Hindered mostly due to high level radiation/contamination.

• Site Restoration:

“Back to Green Field” is practically an impossible goal.– Wide spread contamination: Soil, Groundwater, Seawater

– Highly radiotoxic actinide species (Pu) involved.

– No candidate repository locations available for large volume of heavily contaminated equipment and concrete rubble.

Conclusion: “Plan-A” is not a workable option for Fukushima NPP Units.

71

Unit 1 2 3

mSv/h 46,500 18,200 8,400

Technique Application for Fukushima NPP Units

Implosion

Possible for all buildings other than Rx. Bldg. of Unit 1 to 3 after some decontamination efforts.

Not practical for Rx. Bldg. of Unit 1 to 3 due to high contamination level.

Chemical DecontaminationAlready done for Unit 4 Rx.

Not practical for other units due to too much activity load.

Remove Rx. Internals by High Pressure Abrasive Water Jet

Already done for Unit 4.

Not practical for other units due to high contamination level.

Separate RPV from All Other Connecting Systems

Possible for Unit 4.

Not practical for other units due to harsh radiological environment for workers.

(Drywell Dose Rate as of May 20, 2011.)

Note that the dose limit for Emergency Workers is

250mSv.

72


73

Developing “Plan-B”

• Worst Case Scenario– Historical worst case: Chernobyl Accident– How better is Fukushima NPP’s case relative to Chernobyl?– How worse could Fukushima NPP units have been if cooling

capability was lost immediately upon SBO?– How bad could Fukushima NPP units be if “Cold Shutdown”

is terminated now?• Unit 1/2/3 Reactors• Unit 4 SFP

• Basic Technical Requirements for “Plan-B” • Transition from “Plan-A” to “Plan-B”

74Source: NUREG/CR-6042 Rev.2

Worst Case Scenario• Historical Worst Case: Chernobyl Accident

Fukushima NPP

NISA: 1,700,000Ci

NSC: 1,000,000Ci

75

Chernobyl Accident Facts:

Definitely worst from many aspects!

• Release: 1,760PBq of I-131• Contamination:

– 3 major “Hot Spots”, including one as far as 500km from the site.

– Large restricted areas

confiscated zone, closed zone, permanent control zone, periodic control zone

• Personnel Exposure (out of 400 workers at the site on the day of accident):– 140 persons 1 – 2 Gy

– 55 persons 2 – 4 Gy



76

Radionuclide Releases During Chernobyl Accident

Source: Chernobyl – Ten Years On (OECD/NEA)


1MC

i = 3

7,00

0TB

q

Daily Release During Chernobyl Accident

78

Source:　 OECD/NEA “Chernobyl Ten Years on Radiological and Health Impact – An Assessment by the NEA Committee on Radiation Protection and Public Health” November 1995

Cs-137 Contamination 10 years later

Vicinity of Fukushima NPP

80km

Equivalent dose rate of 555kBq/m2 contamination is 1.8μSv/h or 15.8mSv/y. Blue colored r

egion on land represents dose rate greater than 0.3μSv/h as of 3/19/2011.

500km

79

Access Restriction due to High Level Contamination

80

Worst Case for Fukushima NPP

• In spite of large amount of release, resulting overall impact was much smaller than that of Chernobyl accident. This is believed to be mostly because of wind blowing west to east. The worst case was avoided by a favorable wind direction.

Prediction by WeatherOnline (UK)

81

Radiological impact estimated by various organizations

Japan - estimated cumulative dose in mSv through 3/11/2012

82

83

Worst Case for Fukushima NPP (cont’d)

What if cooling capability was lost immediately upon SBO?

• Much more heat load, resulting in more aggressive propagation of failures/degradations of FP barriers. (Reactor Pressure Vessel, Primary Containment, and even Man-Made Rock. Note that Man-Made Rock is not credited as an FP barrier against atmospheric release, but it does play a role as an FP barrier against groundwater/soil contamination.)

• Earlier and much more release of radionuclides, with greater contribution from short-lived species.

• Much more CCI, resulting in generation and accumulation of more non-condensable and combustible gases forming radionuclide aerosols containing more radiotoxic particles.

84

Unit 1 2 3 4

# of Fuel Assembly in Rx. 400 548 548 0

Electrical Output (MWe) 460 784 784 0

Thermal Output (MWt) 1,380 2,381 2,381 0

Estimated Residual Heat (MWt)1% of rated Thermal Output

14 24 24 0


2 months after

shutdown1 hour after

shutdown

85

Major Degradation of Reactor Pressure Vessel Bottom Head and Core-Concrete Interaction (CCI), Resulting in Significant Amount of Release of Radioactive Aerosol

Pedestal Doorway

Pedestal

H2O,

CO2

H2O,

CO2

H2, CO

Aerosol

AerosolAerosol

Aerosol

86

Pedestal DoorwayPedestal

Aerosol

Aerosol

Aerosol

Aerosol

H2, CO

H2O,

CO2

H2O,

CO2

Beginning of Primary Containment Melt-Through

87

Gross Failure of Primary Containment due to Melt-Down Progression

Aeros

ol

Aeros

ol

Aeros

ol

Aeros

ol

88

Gross Man-Made Rock (Basemat) Melt-Through

Man-made Rock


Various Gases and Debris Generated during CCI


Breakdown of FP Species

91

Worst Case for Fukushima NPP (cont’d)

How bad could Fukushima NPP units be if “Cold Shutdown” is terminated now (70 days after shutdown)?

• Much less heat load, resulting in less aggressive propagation of failures/degradations of FP barriers (Reactor Pressure Vessel and Primary Containment).

• Much less release of radionuclides, with negligible contribution from short-lived species.

• Much less CCI, resulting in generation and accumulation of less non-condensable and combustible gases forming radionuclide aerosols containing less radiotoxic particles.

• Water left on the Drywell floor suppresses CCI. • Dilution of molten core with various metal and non-metal materi

als would lower temperature and reduces CCI.

92

Radioactive Decay after 70 days

Half-Life Remaining

1 day 8.5 x 10-22

2 days 2.9 x 10-11

3 days 9.5 x 10-8

I-131,

Cs-134, Cs-136, Cs-137, Rb-86,

Te-127m,

Ba-140, Sr-89, Sr-90,

Co-58, Co-60, Ru-103, Ru-196,

Am-241, Cm-242, Cm-244, Nb-95, Nd-147, Pr-143, Y-91, Zr-95,

Ce-141, Ce-144, Pu-238, Pu-239, Pu-240, Pu-241

32 species gone, 27 species left.

93

Gross Failure of Primary Containment Melt-Through

This would be very slow even if it does take place at all.

94

Complete Melt-Down through Man-made Rock (Basemat)

This is even more unlikely. 12~

13m

95


96

Basic Technical Requirements for “Plan-B”

• Achievable and Predictable– No optimistic assumptions allowed. (e.g. Equipment inside Drywel

l no longer functionable.)

• Designed for Short-term and Long-term Solution– Simplicity and Passiveness

– Good for 1,000 years with minimum maintenance.

– Maintain reversibility in case application for long-term solution is abandoned in the future.

• Short-term and Long-term Safety/Security– No Recriticality, H2/Steam Explosions

– Minimum release of radioactivity to external environment

• Lowest Cost and Shortest Schedule– Least labor intensive, minimum personnel exposure.

97

Transition from “Plan-A” to “Plan-B”

• Water Cooling to Gas/Air Cooling (Forced Circulation to Natural Circulation)

• Proposed Remedy for Leaky System– Introduce fine glass fiber mixed with SiC/B4C powder to clog leak p

aths, then use some chemical reaction to create precipitants (e.g. Ca3(PO4)2 ) to further reinforce leaking boundary.

– Apply knowledge gained from GSI-191 study.

• Proposed improvement to minimize I-131 airborne inside and outside Rx. Bldg if such an effort is still considered necessary.– Spray TSP (Trisodium Phosphate) solution.

98


99

“Plan-B”

• Liquidation Strategies for Fukushima NPP Reactors and SFPs– Strategy-I and II for Reactors– Strategy-A and B for SFPs

• Water Treatment and Entombment• On-Site Above-Ground Repository• Design Beyond Millennium• Beyond “Liquidation”

100

Liquidation Strategies for Fukushima NPP Reactors and SFPs

General Approach• New concept replacing traditional and costly practice, that is, full

demolition of the facility and eventually returning the entire site to “Green-Field”. With this new approach, most radioactive material is left as is in the original locations so that significant cost/schedule reduction is expected.

• Thoroughly engineered design, not like the ad hoc technique applied for Chernobyl Unit 4 under emergency situation.

• Applicable not only for those plants affected by major reactor accidents, but also for those plants orderly shutdown permanently upon expiration of license as an alternative choice.

• Phased approach to shift cooling strategies as decay heat load decreases as a function of time . Forced circulation (Water to Helium, Helium to Air) initially, eventually followed by natural convection with no active component to drive the system.

101

General Approach (cont’d)

• This new approach is named as IE2-D (Innovative EngineeredIn-Situ Entombment Decommissioning) and comprised of the following general decommissioning processes and specific processes unique to each Strategy described in later sections separately:

– Remove all new fuel assemblies currently stored in the New Fuel Storage Vault.

– Remove any equipment reasonably recyclable.

– All process systems containing water inside are to be drained, filled with N2 gas and isolated from external environment.

– All rooms and compartments are either solidly filled with concrete, or vented to the general area so as not to allow accumulation of combustible gas.

Unit 1 2 3 4

# of New Fuel Assy 100 28 52 204

102

Specific Strategies

Unit Reactor Spent Fuel Pool

1 Strategy-II Strategy-B

2 Strategy-IIStrategy-A

SFP Not Affected

3 Strategy-II Strategy-B

4Strategy-I

Reactor Systems Not Affected

Strategy-B

103

Strategy-I

Status

• Implementable only for Unit 4 because this is the only unit where most reactor systems are apparently left unaffected.

• However, the unit was structurally significantly damaged due to the hydrogen explosion, resulting in losses of OHC and possibly FHM as well.

• As another impact due to the hydrogen explosion, the integrity of secondary containment is currently lost.

Unit 1 Unit 2 Unit 3 Unit 4

104

Status (cont’d)

• The unit was in the middle of outage where a major modification project, namely “Shroud Replacement”, was taking place at the time of accident. The reactor configuration during this particular outage was very different from that during normal outages, specifically;

– Steam Dryer and Moisture Separator were removed from the vessel and stacked together vertically in the Dryer Separator Pit.

– All Control Blades, Fuel Support Castings, Control Rod Guide Tubes, and Incore Monitors were removed from the vessel and temporarily stored in the SFP.

– Many reactor internal components including, Feedwater Spargers, Core Spray Piping and Spargers, Top Guide, Core Shroud, and Core Plate were removed from the vessel and transferred to the Dryer Separator Pit where some of them were sliced into small pieces for disposal.

105

Key Steps

• Inspect the Reactor Bldg. and determine degree of impact.

• Clean up Refueling Floor.

• Restore and re-establish capability of Secondary Containment.

• Move all removed reactor internals currently stored in the Dryer Separator Pit and SFP in an orderly manner back to the vessel.

• Fill the Reactor Pressure Vessel with concrete.

• Drain Reactor Cavity and Dryer Separator Pit. (These emptied pools will be used for the storage of various contaminated equipment and debris for the future.)

• Re-assembly RPV Head, Mirror Insulation, PCV Head.

• Proceed to the general decommissioning processes.

106


107

Strategy-II

• Proposed Unit-by-Unit Application– Mode 1, 2 (Options A, B1, B2, C)

– Mode 3

• Tentative Mode Change Schedule• Proposed System Lineup

Detail plant unique assessment is necessary.


108

• Proposed Unit-by-Unit Application

UnitHeat Load*

(kWt)Operation Mode**

Min. Flow Rate*** (Nm3/h)

1 1,400

1 He, Forced 55,000

2 Air, Forced 15,000

3 Air, Natural 5,000

2/3 2,400

1 He, Forced65,000

(∆T = 150 deg-C)


3 Air, Natural 5,000

*: Residual heat generation as of 5/11/2011.

**: See conceptual illustrations and proposed system lineup for each operation mode.

***: Required flow rate is calculated to limit the outlet temperature within 100 degrees above the inlet temperature unless otherwise noted.

109

Operation Mode 1 2 3

Heat Generation Range (kW) > 700 200 - 700 < 200

Cooling Strategy He/Forced Air/Forced Air/Natural

Estimated Heat Generation as of 5/11/2011

Unit 1 1,400kW

Unit 2 2,400kW

Unit 3 2,400kW

Unit 4 0 N/A

Mode 1

Mode 2 Mode 3

Mode 2 Mode 3

1Y 2Y 10Y

Mode 1 Mode 2

Mode 1

3Y

Mode 3

• Tentative Mode Change Schedule

110

Medium

Thermal Conductivity


Heat Capacity

W/m ･ K Air = 1 J/kg ･ degC

He 0.1663 5.53 5192

H2O (Steam) 0.0241 0.77 2098

H2O (Liquid) 0.582 - 4217

Air 0.0316 1 1012

Helium:

• 200Yen/Nm3

• 140Yen/liter (liquid)

• 0.1248kg/liter

Helium:

• 200Yen/Nm3

• 140Yen/liter (liquid)

• 0.1248kg/liter

Favorable Thermal Characteristic of Helium

111

Helium is a standard cooling medium for high temperature gas reactors.

GT-MHR (Gas Turbine – Module Helium Reactor)

112

A*

B*

To be added

Scrubber/Gas Cooler

Ventilation System

Mode-1/2

Heat Sink Gravel

Flow from Suppression Chamber

to Drywell

*: See “proposed line-up” for system interfaces for A and B for each unit.

Option A

113

Heat Sink Gravel

Factors to be considered for selection:

• High thermal conductivity

• Radiation shielding

• High performance to absorb radioactive gas/particle.

114

Copper Sphere ShellZeolite

Mixing several different constituents may be considered

115

Flow from Suppression Chamber to Drywell

116

Field Assembly of Primary Containment at Browns Ferry Site During Construction Time

117

Unit 1 Core Spray SystemHelium/Air Injection Point

Proposed System Lineup

A

118

Unit 1 Shutdown Cooling System

B

This valve may not be opened.

119

Unit 1 Isolation Condenser (Alternative Option)

B

X

120

Unit 2/3 Core Spray SystemHelium Injection Point

A

121

Unit 2/3 High Pressure Injection System

B

122

Unit 1 Atmospheric Control System (Alternative Option)

B

X

123

To be added

To be added

Scrubber/Gas Cooler

Ventilation System

Rx. Bldg. Truck Bay

Mode-1/2

Option B1

124

To be added

To be added

Scrubber/Gas Cooler

Ventilation System

Rx. Bldg. Truck Bay

Blower

Mode-1/2

Option B2

125

To be added

To be added

Scrubber/Gas Cooler

Ventilation System

Rx. Bldg. Truck Bay

Mode-2

Option C

126

Rx. Bldg. Truck Bay

StackAir Gap for Flow Path

Air Flow only by Natural Convection

Mode-3

See detail “D”

See detail “E”

127

Detail “D”

128

Detail “D”

129

Construction Details of Bottom Portion of Primary Containment Vessel (Oyster Creek) (2)

Detail “E”

130

Construction Details of Bottom Portion of Primary Containment Vessel (Oyster Creek) (1)

Detail “E”

131


132

Strategy-A

• Implementable only for Unit 2 because both OHC and FHM are seemingly still functionable and available.

Key Steps:• Inspect (sipping and visual examination) on all fuel assemblies

and identify any damage fuel.• Load only undamaged/non-degraded fuel assemblies into cask

s for:– Wet Storage at Common Storage Pool or other designated site(s).

– Reprocessing for MOX at Rokkasho facility.

– Dry Storage at site or other designated site(s).


133

• Damaged/degraded fuel assemblies are treated differently.– No certified cask design currently available.

– Design and certify special cask only for this group of fuel and transfer to other unit (1, 3, or 4) for Strategy-B.

– Leave only this group of fuel at Unit 2 and apply Strategy-B.

134


135

Strategy-B

• Proposed Unit-by-Unit Application– Mode 1, 2, 3

• Comparison with “Plan-A”

Advantages vs. Disadvantages– General Comparison– Cost– Schedule– Security during implementation


136

• Proposed Unit-by-Unit Application

UnitHeat Generation

(kWt)Recommended

Operation Mode*Min. Flow Rate**

(Nm3/h)

1 70 3 Air Natural 1,930

2 4601 He***, Forced 18,000


3 2302 Air, Forced 6,500

3 Air Natural 6,500

4 1,800 1 He***, Forced 70,000

*: See later section for the definition of each operation mode.

**: Required flow rate is calculated to limit the outlet temperature within 100 degrees above the inlet temperature.

***: He is more recommendable because of its higher heat conductivity and lower viscosity (flow friction).

137

Operation Mode 1 2 3

Heat Generation Range (kW) > 350 100 - 350 < 100

Cooling Strategy He/Forced Air/Forced Air/Natural

Estimated Heat Generation as of 5/11/2011

Unit 1 70kW

Unit 2 460kW

Unit 3 230kW

Unit 4 1,800kW Mode 3Mode 2Mode 1

Mode 1

Mode 2 Mode 3

Mode 3

Mode 2 Mode 3

0.5Y 2Y 5Y 5.5Y 10Y

• Tentative Mode Change Schedule

138

N

Spent Fuel Racks Spent Fuel Racks

Gate

Cask Pit

A A

Spent Fuel Pool (top view)

139

Water Level

Spent Fuel Racks Spent Fuel Racks

Spent Fuel Pool (side view)

A-A

140

Finned Heat Sink Chambers (Copper)

Step-1 Install Finned Heat Sink Chambers on Spent Fuel Racks.

141

Cross-Tie Pipes

142

Step-2 Install Pre-fabricated Pipe Modules.

143

35cm

Cold (Inlet) 2-inch Sch#40 Stainless Steel

Hot (Outlet) 2-inch Sch#40 Stainless Steel

Convection Cooling 2-inch Copper

Pipe Modules

144

eachcm,xcmAvailableAreaFlowTotal

each.approxInstalledTubesofNumber

m)SurfaceTopEntireof%(~AvailableAreaTotalAssumed

cmCellUnitofArea

22

2

22

9001890021

900

5050

530354

3

Φ10mm　　　 (typ. 4)

50m

mA

pp

rox.

200

0mm

Ap

pro

x. 8

000m

m

145

A A

View A-A

Main Header

Main Header

Main Header

Distribution Header

Top View

146

147

Step-3 Load Heat Sink Gravel

Water

Gravel

148

Heat Sink Gravel

Factors to be considered:

• High thermal conductivity

• Radiation shielding

• High performance to absorb radioactive gas/particle.

149

Copper Sphere ShellZeolite

Mixing several different constituents may be considered

150

Water level gradually decreases

Step-4 Start Ventilation System

151

To Ventilation Fan and Gas Treatment System

Wet Scrubber

Water level

Operation Mode 1, and Mode 2

152

Mode Heat Load Cooling MediumCleanup System

Power

1High

(>350kW)

Contaminated Helium

Required Forced Cooling

2Medium

(100-350kW)

Contaminated Air

Required Forced Cooling

3Low

(<100kW)

Non-contaminated Air

Not RequiredNatural

Convection

Operation Mode

Medium


Relative Thermal Conductivity

Heat Capacity

W/m ･ K Air = 1 J/kg ･ K

He 0.1663 5.53 5192

Air 0.0316 1 1012

Favorable Thermal Characteristic of Helium

153

Operation Mode 3 “Natural Convection”

Inlet Sleeve

Shielded Air Intake Block

154

155

Comparison with “Plan-A”

Advantages vs. Disadvantages• General Comparison

– Favorable for “Plan-B”

“Plan-A” “Plan-B” Strategy-B

FHMRequired but currently not available due to damage caused by H2 explosion.

Not required.

OHC Ditto Not required

Fuel Inspection (Sipping)

Required Not required.

Spent Fuel StorageNo certified Transportation/Storage Cask for damaged fuel assemblies currently available.

Not required.

ISFSI ?? Not required

SecurityCurrently exposed to very poor conditions.

Duration of poor security conditions can be minimized.

156

Advantages vs. Disadvantages (cont’d)• General Comparison

– Potential issues associated with “Plan-B” Strategy-B.

“Plan-A” “Plan-B” Strategy-B

Geological DisposalCan be eventually transferred to this option.

• Practice not pursued previously.• Buried under man-made structure significantly above ground elevation.

Licensing ProcessRelatively more predictable.

• Unknown. No previous experience. • No siting criteria established.

Safety AnalysisRelatively more predictable.

Various supporting analysis necessary.• Design Basis Accident (DBA)

Security Issue Currently very poor.Permanent measures including Aircraft Impact Assessment (AIA) necessary.

Public Acceptance Unknown. Unknown.

157

• Cost/Schedule– Cost/schedule potentially eliminated by applying “Plan-B”

Strategy-B.

– Cost for “Plan-B” Strategy-B: much less than that for restoring OHC alone.

– Schedule for “Plan-B” Strategy-B: much shorter than that for unloading spent fuel from SFP alone.

Activity Cost (JPY) Schedule (Year)

Restore OHC X billion ~2

Restore FHM

Fuel Inspection (Sipping)

Procure Spent Fuel Casks X billion

Unload Spent Fuel from SFP ~5

Construct ISFSI

158

Conclusions:• Practical approach for Units 1, 3 and 4. (Strategy-A is considered

implementable only for Unit 2.)

• Advantage of “Plan-B” Strategy-B over “Plan-A” is obvious.

• All associated technical issues are manageable.

• Two potentially challenging non-technical issues:

– Licensing

– Public Acceptance

159

Ultimate Configuration with Operation Mode 3

All contaminated equipment and materials are permanently buried in-situ.

Paradigm Shift !!

This concept, in spite of huge cost benefit expected, significantly deviates from the conventional approach.

Paradigm Shift !!

This concept, in spite of huge cost benefit expected, significantly deviates from the conventional approach.

160


161

Water Treatment and Entombment

• Water Treatment System is a part of “Plan-B” and integrated into IE2-D strategies.– Low level contaminated water is used as a water source to

produce ready-mixed concrete for general purpose.– Highly concentrated radioactive water is vitrified (because

of relatively high heat generation) and stored at On-Site Repository.

• Water to be processed:– Highly contaminated water currently stored in various pools

at site.– Contaminated sea water within Intake Area.

162

Water Treatment (1)

Highly Contaminated Water Currently

Stored in Various Pools at Site

Vitrification

Canisters

On-site Repository

Concentrated Radioactive Liquid

Treatment System

Cement

Aggregate

Contaminated Concrete Rubble (Optional)

Processed Water (slight contamination

allowed)

Ready-Mixed Concrete

< 5,000Bq/cm3

< 0.065mSv/h

20 v/v% 80 v/v%

For Entombment Work

For Entombment Work

163

Dose Rate Calculation of Homogenously Contaminated Concrete

h/mSv.or,h/SvR

hBq/Svx.

e.

R

m/BqQ

eQ

dreQR

m.r

er

QdrrdR

.x.

.r. r

r

065065

10928

12517

104

10

44

828

14

8

82825179

39

828

0

828

0

22

Assumption:

500TBq in 105 m3, or 5,000Bq/cm3 of processed water

Water Content in Ready Mixed Concrete = 20%

Calculation:

Low enough!

164

Heat Generation Calculation of Homogenously Contaminated Concrete

Assumption:

500TBq in 105 m3 of water, or 5,000Bq/cm3

Water Content in Ready Mixed Concrete = 20%

Energy Release per Disintegration = 1MeV

Calculation:

Total energy release rate = (1.6 x 10-13J) x (5 x 1014/sec) = 80W

Temperature increase based on black body radiation

q” = σT4 σ = 5.67 x 10-8

q” = 80/(4πr2) r = 28.8m

q” = 7.7 x 10-3 W/m2

T = 19-deg C

Low enough!

165

Water Treatment (2)

Desalination System

Cement

Aggregate

Contaminated Concrete Rubble (Optional)

Processed Water (still slightly

contaminated)

Ready-Mixed Concrete

< 5,000Bq/cm3

< 0.065mSv/h

20 v/v% 80 v/v%

Contaminated Water within Intake Area

For Encasing Concrete Rubble

For Encasing Concrete Rubble

166


167

On-Site Above-Ground Repository

• New site arrangement consists of three major islands, each enclosed by an individual protected area:– ISFSI– On-Site Repository for vitrified canisters– IE2-Ded Reactors

168

Intake Area

Intake Facility (typ.)

Backwash Valve Pit (typ.)

Control Bldg. (typ.) Turbine

Bldg. (typ.)

Rx. Bldg. (typ.)

RW Bldg. (typ.)

BeforeUnit 2

Unit 1Unit 3Unit 4

169

Entombed ReactorsAfter

Protection Fence against Aircraft Impact

Stack

Concrete rubble generated from demolition of all other structures is encased in the large concrete block(s).

Tsunami Barrier

Tsunami Barrier

Wave Breakers for Tsunami Protection

Original Shoreline

170

ISFSI for SNF and any potential GTCC Waste

On-Site Repository for Vitrified Canisters

Legend:

Monitoring Post

Ground Water Sampling Point

Protected Area

Main Gate

New Site Boundary

Monitoring Facility

Conceptual New Site Arrangement

171


172

Design Beyond Millennium

• IE2-Ded Reactors, ISFSI, and On-Site Repository must be qualified for long-term endurance.

• Traditionally, man-made structures were not credited for this purpose.

173

But, man-made structures may not be too bad…

Possibly good for centuries or even millennia!

174

Source: “The Future of Nuclear Power” (MIT)

Residual Heat

1/20

175

Source: “The Future of Nuclear Power” (MIT)

Radioactivity

1/100

176


177

District for New Industry/Community Development

Entombed Reactors (Units 1 to 4)

Survived Reactors　　(Units 5 and 6)

Solar Thermal PowerBeyond “Liquidation”

Previous Site Boundary

178

Target Overall Schedule

Activities 2y 4y 6y 8y 10y

Public Acceptance (Workshop)

Licensing Review on EI2-D (Safety Analysis)

Build Liquidators’ Villages

Recruit Liquidators

Expand On-Site Liquidation Infrastructures

EI2-D Projects

Unit 1 to 3, Reactor

Unit 4, Reactor

Unit 1, 3, and 4, SFP

Unit 2, SFP

Demolition of other structures

Construct On-Site Repository Facility

Water Treatment, Vitrification

Construct Intake Area Tsunami Barriers

Construct ISFSI (for Unit 2 SNF)

New Industry/Community Development

FS, Bidding, Design/EngineeringMode 1 Mode 2 Mode 3

Mode 1 Mode 2 Mode 3

Sipping Transportation Campaign

179


180

Next Step

• Feasibility study by independent organization(s):– Technical aspect

– Financial aspect

– Political aspect

• Survey on public opinions.• Voices from international communities.

• Issues:– Interactions with decision-makers

– Financial support to proceed to design phase

– How to make a go-no-go decision

181

Analysis for Future Benefit

• Better predictability and versatility

182

Time Cooling Efforts Abandoned after Plant Shutdown

Release (NG, I, Cs)

CCI Penetration Depth

Required Evacuation Radius

183

184

Temperature Monitoring Probes

185

To be added

Rx. Bldg. Truck Bay

186

187

Technology

Plan B